1. Field of the Invention
The present invention generally relates to the shape and disposition of an internal structural member constituting part of an internal structure and disposed within a pressurized water reactor in which heat exchange takes place between heating members and a coolant. More particularly, the invention is concerned with the shape and disposition of coolant flow stabilizing members disposed within an upper plenum in the vicinity of outlet nozzles for stabilizing the coolant flow in regions located close to the outlet nozzles.
2. Description of the Related Art
In a pressurized water reactor for a power plant, light water serving as a coolant is fed to a reactor core to be heated to a high temperature by heat generated through nuclear fission. The heated light water is taken out and supplied to a steam generator for generating steam which is then fed to a steam system provided separately from a core cooling system to be utilized for rotationally driving a steam turbine and a rotor of an electric generator operatively coupled to the turbine to thereby generate electric energy.
For a better understanding of the invention, background techniques thereof will be reviewed below. FIG. 8 of the accompanying drawings shows an internal structure of a pressurized water reactor typical of nuclear reactors. Referring to the figure, accommodated within a nuclear reactor vessel 10 are reactor core internals, nuclear fuel assemblies, control rods, control rod cluster guide tubes, support members and others. Described in brief, the nuclear reactor vessel 10 is integrally provided with inlet nozzles 11 and outlet nozzles 12 for the reactor coolant which is light water, and a core barrel 30 is suspended vertically within the nuclear reactor vessel 10. The number of the inlet nozzles 11 and the outlet nozzles 12, respectively, coincide with the number of coolant circulation loops which in turn depends on the output power rating of the reactor. Ordinarily, the number of the inlet nozzles 11 and outlet nozzles 12 ranges from two to four each.
By way of example, in a power plant with a large power generation capacity, there are ordinarily provided a plurality of coolant circulation loops in consideration of the capacities of pumps and the steam generator employed in the cooling system as well as restrictions imposed with respect to the disposition thereof within a containment vessel. In other words, the a number of coolant circulation loops installed is determined depending on the output capacity. Since a reactor power plant with a large output capacity is provided with four coolant circulation loops, the number of the inlet nozzles 11 and the outlet nozzles 12 is also four, respectively. The inlet nozzles 11 and the outlet nozzles 12 are installed in the nuclear reactor vessel at predetermined intervals in the circumferential direction thereof. Further, a lower core support plate 32 and a lower core plate 31 are horizontally disposed within the core barrel 30 at a lower portion thereof, respectively. A bottom or lower plenum 41 is defined beneath the lower core support plate 32.
Mounted on the lower core plate 31 are a large number of loaded fuel assemblies 33 which are disposed adjacent to one another to thereby constitute a reactor core. Disposed above the fuel assemblies 33 is an upper core plate 21 which is supported by an upper core support plate 20 by way of upper core support columns 23. The fuel assemblies 33 are pressed downwards by means of the upper core plate 21 so that the fuel assemblies 33 are prevented from being displaced upwards under the influence of buoyancy exerted by the flowing coolant. A plurality of control rod cluster guide tubes 22 are supported at lower ends thereof on the upper surface of the upper core plate 21 by means of supporting pins or the like (not shown). The control rod cluster guide tubes 22 extend upwardly through and beyond the upper core support plate 20. By withdrawing the control rod clusters (not shown either) from the reactor core through the medium of the control rod cluster guide tubes 22 or inserting the control rod clusters into the fuel assemblies 33 of the reactor core through the control rod cluster guide tubes 22, the thermal output of-the reactor core can be adjusted.
The upper core plate 21 and the upper core support plate 20 are interconnected by means of the upper core support columns 23 in order to ensure structurally high strength or robustness. Further, the control rod cluster guide tubes 22 extending through the upper core support plate 20 are fixedly secured to the upper core support plate 20. Thus, the control rod cluster guide tubes 22 are protected against displacement or dislocation in a lateral or transverse direction. Defined between the upper core plate 21 and the upper core support plate 20 interconnected as mentioned above is an upper plenum 40.
Next, description will be directed to the flow or streams of light water employed as the coolant within the nuclear reactor vessel 10 realized in the structure described above. Referring to FIG. 8, light water of low temperature fed to the nuclear reactor vessel 10 by way of the inlet nozzles 11 flows as indicated by the arrows in FIG. 8. More specifically, light water fed to the nuclear reactor vessel 10 through the inlet nozzle 11 flows at first downwardly through an annular space defined between the outer surface of the core barrel 30 and the inner wall of the nuclear reactor vessel 10. The light water is forced to turn upwards within the lower plenum 41. Thereafter, the light water flows into the reactor core after passing through the lower core support plate 32 and the lower core plate 31. Within the reactor core, light water flows upwardly substantially in parallel. In the course of flowing through the reactor core, heat generated by the fuel rods of the fuel assemblies is absorbed by the light water, which results in a temperature increase thereof. After passing through the upper core plate 21, the flowing direction of light water changes to a horizontal or transverse direction. Finally, light water leaves the nuclear reactor vessel 10 through the outlet nozzle 12 to be supplied to a steam generator (not shown) by way of an outlet pipe 42.
More specifically, within the upper plenum 40 defined above the upper core plate 21, light water flows radially outward from a central region of the core to reach the inner wall of the core barrel 30 whereupon the light water flows toward the outlet nozzle 12 along the inner wall of the core barrel 30 in a space surrounding the outer periphery of the core. A portion of the light water flows in one direction along the inner wall of the core barrel 30 while another portion of the light water flows in the other direction opposite to the previous direction along the inner wall of the core barrel 30. Thus, in a space located within the upper plenum near but bellow the outlet nozzle 12, between the streams of light water flowing in opposite directions collide. After the collision, the flow directions of the light water are changed so as to flow upward toward the outlet nozzle 12. Such being the case, the flow of light water in the space located in the vicinity of the outlet nozzle 12 is made unstable due to the above mentioned collision as well as occurrence of turbulence such as swirls or vortexes.
Also, the nuclear reactor vessel 10 of the four-loop reactor plant mentioned previously is implemented in a structure such that two outlet nozzles 12 are disposed adjacent to each other because of the restrictions imposed in view of the requirement for realizing the structure of the primary coolant loop and the reactor containment vessel in a reduced size and the like, as can be seen in FIGS. 9 and 10 of the accompanying drawings. Consequently, some of the light water flowing along the inner wall of the core barrel 30 tends to flow transversely beneath the outlets of the outlet nozzles 12, and will also collide with the coolant streams in a region between the adjacent outlet nozzles 12. After the collision, light water is forced to change its flow directions so as to flow upwardly. However, due to occurrence of turbulence such as swirls or vortexes and differences in the flow rate between the two loops, light water can rarely flow uniformly into two loops through the adjacent outlet nozzles 12, respectively. Thus, an unstable water stagnation region unavoidably occurs in the vicinity of the outlet nozzles 12. As a result of this, the behavior of light water flowing into the outlet nozzles 12 becomes complicated and unstable.
It should further be added that under the influence of temperature distribution in the fuel region, a stratified flow distribution within the upper plenum 40 is formed such that the coolant of relatively high temperature flows at a relatively high level while the coolant of relatively low temperature flows at a relatively low level. In other words, a plurality of stratified coolant flows of different temperatures are produced within the upper plenum 40, as viewed in the vertical direction. Such being the case, the coolant flowing on or in proximity to the upper surface of the upper core plate 21 along the inner wall of the core barrel 30 is at a relatively low temperature.
For the reasons mentioned above, the unstable coolant flow of a relatively low temperature brings about fluctuations or variations in the temperature of the coolant flowing through the outlet pipe 42 connected to the outlet nozzle 12. Consequently, this can easily become an obstacle when mean temperature of the nuclear reactor is measured by measuring the temperature of the coolant flowing through the outlet pipe 42 connected to the outlet nozzle 12.
In light of the state of the art described above, it is an object of the present invention to provide an internal structure disposed within a nuclear reactor vessel, the structure including an internal structural member configured and disposed so as to be capable of stabilizing the coolant flow in a region located near or close to an outlet nozzle, to thereby eliminate or mitigate difficulties which may otherwise be involved in the temperature measurement of the coolant flowing through the outlet pipe.
In view of the above and other objects which will become apparent as the description proceeds, the present invention is directed to an internal structure of a nuclear reactor including an upper plenum which is defined above a fuel region through which a heated coolant flows and which is hydraulically communicated with a plurality of outlet nozzles mounted on a side wall of a nuclear reactor vessel.
According to a general aspect of the present invention, it is proposed that in the internal structure of the nuclear reactor mentioned above, at least a short flow stabilizing member which is lower than the outlet nozzle is disposed in the vicinity of the core barrel in a region outside of the fuel region.
By virtue of the arrangement described above, the currents or streams of the heated coolant flowing in the vicinity of the core barrel at a level lower than the outlet nozzle within the upper plenum are smoothly guided upwards by the short flow stabilizing member without undergoing collision. Thus, unstable coolant flow in the vicinity of the outlet nozzle can be avoided. Accordingly, temperature fluctuation of the heated coolant flowing through the outlet pipe connected to the outlet nozzle can be suppressed, whereby the average or mean temperature measurement mentioned previously can be carried out with enhanced stability and reliability. Moreover, since the short flow stabilizing member is lower than the outlet nozzle, the hydrodynamic load acting on the flow stabilizing member can be reduced. Thus, the soundness of the reactor internal structure can be ensured.
In a preferred mode for carrying out the present invention, the short flow stabilizing member may be disposed in the vicinity of an outlet of the outlet nozzle such that a top end of the short flow stabilizing member is positioned beneath the outlet of the outlet nozzle. By virtue of this arrangement, the stream or currents of the heated coolant flowing in the vicinity of the core barrel at a lower level within the upper plenum are guided by the flow stabilizing member toward the outlet nozzle from the lower side thereof, whereby the flow of the heated coolant entering the outlet nozzle can further be stabilized.
In another preferred mode for carrying out the invention, the short flow stabilizing member may be disposed at a position directly underneath a center portion of the outlet of the outlet nozzle. With such an arrangement, the stream of the heated coolant flowing within the upper plenum in the vicinity of the core barrel at a lower level can be guided uniformly into the outlet nozzle at both sides of the short flow stabilizing member without colliding. In this way, the flow of the heated coolant can further be stabilized.
Further, the teachings of the present invention can equally be applied to a nuclear reactor for a four-loop plant as well. In that case, the short flow stabilizing members may be disposed at positions substantially midway between two adjacent outlet nozzles. With the arrangement in which the short flow stabilizing members are disposed between the adjacent outlet nozzles in the reactor for a four-loop plant, the streams of the heated coolant flowing within the upper plenum in the vicinity of the core barrel at a lower level can be prevented from colliding, and thus the coolant can flow into the outlet nozzles of every loop with enhanced stability.
Moreover, it should be added that the length of the flow stabilizing member should preferably be selected when the mounted, the top end thereof is positioned at a height level within a range of from a position midway between the upper surface of the upper core plate and the lowermost portion of the outlet of the outlet nozzle to a position level lower than the lowermost portion of the outlet. With the dimensions of the short flow stabilizing member mentioned above, there is essentially no possibility that the heated coolant flowing toward the outlet nozzle will be blocked. Moreover, a hydrodynamic load acting on the short flow stabilizing member can be reduced. Thus, the soundness of the internal structure of the reactor can be ensured. Further, because of the short length of the flow stabilizing member, handling for mounting can be facilitated. Thus, the short flow stabilizing member can be employed in existing nuclear reactors as well.